Manufacturing of zirconia microspheres doped with erbia, yttria and ceria by internal gelation process as a part of a cermet fuel

Zirconium oxide is an inert matrix candidate for the transmutation of plutonium in light water reactor (LWR). The thermal conductivity of cubic zirconia is however lower than the conductivities of UO2 and MOX. Special designs are therefore necessary to avoid high peaking temperatures close to the melting point in the zirconia pellet. Cermet would be a favorable design to improve the thermal conductivity. The suggested cermet fuel consists of fine plutonium doped stabilized zirconia particles dispersed in a metallic inert matrix. Manufacturing tests on cubic zirconia microspheres were carried out by using the internal gelation process developed at the Paul Scherrer Institute. Gelation was conducted successfully and the sintered spheres had a homogeneous single cubic structure. The lattice parameter of the cubic zirconia was estimated as a function of the Er, Y and Ce atomic fraction using a simplified semi-quantitative formula. On the experimental side, it is necessary to further investigate the ideal fabrication conditions, because some gel spheres were opaque and fragile and most of the sintered spheres were cracked, nicked and porous

高レベル放射性廃液ガラス固化体の長期健全性に関する研究

第1章 序論 第2章 α照射によるガラス固化体の体積変化に関する研究 第3章 α照射によるガラス固化体の微細構造の変化に関する研究 第4章 α照射によるガラス固化体の機械的性質の変化に関する研究 第5章 脱イオン水中でのガラス固化体の浸出挙動に関する研究 第6章 緩衝材共存下でのガラス固化体の浸出挙動に関する研究 第7章 結論主1-参1工学_エネルギー量子_応用原子

Glass dissolution rate measurement and calculation revisited

International audienc

High burn-up operation and MOX burning in LWR; Effects of burn-up and extended cooling period of spent fuel on vitrification and disposal

<p>Looking ahead to final disposal of high-level radioactive waste arising from further utilization of nuclear energy, the effects of high burn-up of light-water reactors (LWR) with UO₂ and MOX fuel and extended cooling period of spent fuel on waste management and disposal were discussed. It was assumed that the waste loading of waste glass is restricted by three factors: heat generation rate, MoO₃ content, and platinum group metal content. As a result of evaluation for effects of extended cooling period, the waste loading of waste glass from both UO₂ and MOX spent fuel could be increased in the current vitrification technology. For the storage of waste glass from MOX spent fuel with higher waste loading, however, those waste glass require long storage period prior to geological disposal because decay heat of ²⁴¹Am contributes significantly. Therefore, the evaluation of effects of Am separation on the storage period was performed. Furthermore, heat transfer calculation was carried out in order to evaluate the temperature of buffer material in a geological repository. The results showed, 70 to 90% of Am separation is sufficiently effective in terms of thermal feasibility of a repository.</p